SBK-L-12258, LRA Commitment Update and RAI B. 1.4-4 Clarification, NextEra Energy Seabrook License Renewal Application

LRA Commitment Update and RAI B. 1.4-4 Clarification, NextEra Energy Seabrook License Renewal Application

ML12349A214
Person / Time
Site:	Seabrook
Issue date:	12/10/2012
From:	Walsh K NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-12258
Download: ML12349A214 (25)

Text

NEXT~Rf ENBERGY741 December 10, 2012 SBK-L-12258 Docket No. 50-443 U.S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station LRA Commitment Update and RAI B. 1.4-4 Clarification NextEra Energy Seabrook License Renewal Application

References:

1. NextEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010. (Accession Number ML101590099)

2. NextEra Energy Seabrook, LLC letter SBK-L-12084, "Seabrook Station Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Supplemental Response - RAI B.2.1.11-2 and B.2.1.12-6," April 26, 2012.

(Accession Number ML121220298)

3. NextEra Energy Seabrook, LLC letter SBK-L-12123, "Seabrook Station Next Era Energy Seabrook License Renewal Application Supplement #25," June 19, 2012. (Accession Number ML12178A405)

4. NextEra Energy Seabrook, LLC letter SBK-L-12072, "Request for Permanent Application of Steam Generator Tube Alternate Repair Criteria, H*," April 10, 2012. (Accession Number ML12121A527)

5. Amendment 131 to facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1, "Seabrook Station, Unit No.1 - Issuance of Amendment Re: Permanent Application of Steam Generator Tube Alternate repair Criteria, H*," September 10, 2012.

(Accession Number ML12178A537)

6. NextEra Energy Seabrook, LLC letter SBK-L-12183, "Seabrook Station Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Request for Additional Information - Set 18 (Operating Experience),"

September 18, 2012. (Accession Number ML12268A170)

In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

In Reference 2, NextEra last confirmed the scheduled date for implementation of commitment number 52, "Implement measures to maintain the exterior surface of the Containment Structure, from elevation -30 feet to +20 feet dewatered by December 31, 2012." Measures have been put NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 f* )ULJ

United States Nuclear Regulatory Commission SBK-L-12258 / Page 2 in place to maintain the subject area dewatered. LRA Table A.3 has been revised to reflect commitment number 52 as ongoing. This change is reflected in the revised commitment list included in Enclosure 3.

In Reference 3, NextEra responded to the NRC requests for additional information related to removal of a seal cap enclosure from SI-V-82, where it was stated that the removal of the seal cap enclosure and restoration of the original configuration would be completed no later than December 31, 2014. The valve, including its seal cap enclosure, was removed and the valve replaced during the recent OR15 reflieling outage. This item is complete. Enclosure 1 contains changes to the LRA associated with completion of this commitment.

Also in Reference 3, NextEra provided clarification to the response provided for RAI B.2.1.10-1 related to Steam Generator Tube-to-Tubesheet Weld Inspection Plans. In that letter, commitment number 54 was revised to address the potential for cracking due to PWSCC of tube-to-tubesheet welds using one of two options specified. The evaluation specified in commitment number 54, option #2, is complete (Reference 4). Permanent application of the steam generator tube alternate repair criteria, H* is documented in Amendment No. 131 to the Facility Operating License No. NPF-86 (Reference 5). This item is complete. Enclosure 1 contains changes to the LRA associated with completion of this commitment. This change is also reflected in the revised commitment list included in Enclosure 3.

In Reference 6, NextEra responded to requests for additional information related to operating experience. During a telephone conference with the NRC Staff on December 4, 2012, clarification was requested pertaining to the response to RAI B. 1.4-4 c), as to the responsibilities of the License Renewal Engineer, and sharing of site operating experience with the rest of the nuclear industry. This clarification is provided in Enclosure 2.

Provided in this Supplement are changes to the License Renewal Application (LRA). To facilitate understanding, the changes are explained, and where appropriate, portions of the LRA are repeated with the change highlighted by strikethroughs for deleted text and bolded italics for inserted text.

There are no new or revised regulatory commitments contained in this letter, however, completion status of previously made commitments has changed. Enclosure 3 contains an update to LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List, to reflect completion status changes.

If there are any questions or additional information is needed, please contact Mr. Richard R.

Cliche, License Renewal Project Manager, at (603) 773-7003.

If you have any questions regarding this correspondence, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.

Sincerely, NextEra FA SaokLLC.

Kevin T. Walsh Site Vice President

United States Nuclear Regulatory Commission SBK-L-12258 / Page 3 - Changes to the LRA Associated with Completed Commitments - Clarification of Response to RAI B. 1.4-4 c) - LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Completion Status Changes.

cc:

W.M. Dean, NRC Region I Administrator J. G. Lamb, NRC Project Manager, Project Directorate 1-2 S. Rich, NRC Senior Resident Inspector P. D. Milano, NRC Project Manager, License Renewal L. M. James, NRC Project Manager, License Renewal Mr. Christopher M. Pope Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

United States Nuclear Regulatory Commission SBK-L-12258 / Page 4 NEXTera BRO0 0K

ýSEA I, Kevin Walsh, Site Vice President of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed Before me this

/ZO day of December, 2012 Kevin T. Walsh Site Vice President Notary Pu/ic

Enclosure 1 to SBK-L-12258 Changes to the LRA Associated with Completed Commitments

United States Nuclear Regulatory Commission Page 2 of 6 SBK-L-12258 / Enclosure 1 Commitment: Removal of the seal cap enclosures and restoration of the original configuration to be completed no later than December 31, 2014.

In NextEra Energy Seabrook, LLC letter SBK-L-12123 (Reference 3), NextEra responded to the NRC request for additional information RAI B.2.1.9-2 related to removal of a seal cap enclosure from SI-V-82, where it was stated that the removal of the seal cap enclosure and restoration of the original configuration would be completed no later than December 31, 2014.

The valve, including its seal cap enclosure, was removed and the valve replaced during the recent OR15 refueling outage. This item has been completed. As a result, the following changes are made to the LRA.

1) License Renewal Application Appendix A, Section A.2.1.9, page A-9, is revised by deleting the second paragraph added by Reference 3 as follows.

Seabrook Station has one seal cap enelosurfe that suf~ounds the pressur-e retaining bolts oa valve 1 SI V 82. The seal cap enelesures on S! V 82 was installed during the FLreed Outage to apllw cationed operation o.f the unit until such time that the valve eould be repaired. The installation of a seal cap enclosures cr-eates a submierged enivir-onffenit that prevents the aging management of the belting and component extelal surfaces for-loss of material, loss of preload, craeldng, anad change in material propeffies. Therefeore, removal of the seal cap encloT.ses and restoration of the original e.nfiguration is plauled to ompleed be no latrtan Dccembcr 31, 2014. With the remoeval of the seal cap enclo.........Su.S the existi..n a ... anagement programs will remain sufficient to age manage the bolting and component extcrnal sur-faccs for loss of material, loss of pr-eload-,

efaeking, and change in material propeffies during the period of extended o~perationt.

2) License Renewal Application Appendix B, Section B.2. 1.9, page B-60, is revised by deleting the second paragraph added by Reference 3 as follows.

Seabrook Station has onie seal cap enclostifes that suffouinds the pressurfe retaininig bolts of-valve 1 S! V 82. The seal cap enclosurfes on SI V 82 was installed durfing the 2011 For-eed Outage to allow continued operationt of the unit until suceh time that the valve couild be r-epaired. The installation of a seal cap enclosurfes creates a submerged envirnment tha prevents'he aging maniagement of the bolting and component externa surfacies for-loss of material, loss of preload, cracking, wand change in material properties.

Therefore, removal of the seal cap enclosurfes and r-estoration of the original configuratio is planned to be completed no later than December 31, 2011. With the removal of the seal cap enclosurfes the existining ngmngement programs will remain suifficient to aeage ge, theboling- and componenit external surfactee for los of material, loseo prelead, craekldng, and change in material properties durfing the period of extende opera4ien.

3) License Renewal Application Appendix B, Section B.2.1.9, page B-57, is revised by adding new Operating Experience item 6 as follows.

6. NRC Information Notice 2012-15 was issued in August of 2012. This notice was issued to inform addressees of potential issues associated with the installation of seal cap enclosures (enclosures)to mitigate leakagefrom A-286 bolted connections

United States Nuclear Regulatory Commission Page 3 of 6 SBK-L-12258 / Enclosure 1 in nuclearpower plantpiping. During a 2011 ForcedOutage Seabrook, installeda seal cap enclosure on a bolted bonnet check valve to allow continued operation of the unit until such time that the valve could be repaired. The installationof a seal cap enclosure created a submerged environment that would prevent the aging management of the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in materialproperties. In June of 2011, Seabrook Station notified the NRC License Renewal Staff of this condition and stated plans to remove the seal cap enclosure no later than December 31, 2014.

The valve, including its seal cap enclosure, was removed and the valve replaced in the fall of 2012, during Refuel Outage 15. With the removal of the seal cap enclosure, the existing aging management programs remain sufficient to age manage the bolting and component external surfaces for loss of material, loss of preload, cracking, and change in materialproperties during the period of extended operation.

United States Nuclear Regulatory Commission Page 4 of 6 SBK-L-12258 / Enclosure 1 Commitment (#54): Address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two specified options at least 24 months prior to entering the period of extended operation.

In NextEra Energy Seabrook, LLC letter SBK-L-12123 (Reference 3), clarification was provided to RAI B.2. 1.10-1 response related to the Steam Generator Tube Integrity Program. A commitment (#54) was made in that letter to address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of two specified options at least 24 months prior to entering the period of extended operation:.

Amendment 131 to the Facility Operating License No. NPF-86 for Seabrook Station, Unit No. 1 was issued on September 10, 2012 (ML12178A537). This amendment provides permanent application of steam generator tube alternate repair criteria, H*, thus satisfying completion of LRA commitment #54. Based on this amendment, NextEra has revised the response provided in Reference 3 to RAI B.2:l.10-1 as follows:.

NextEra Energy Seabrook Revised Response to RAI B.2. 1.10- 1:

1. Based on the eufentlvy aenroved altcate r-epair criteria, the Seabr,.ok Station steam gencrator tube to tubesheet welds are not includcd in the rcactor coolant pressure boundary This alternate repair criteria has net yet bee p.......ly approved. Amendment 131 to the Facility Operating License No. NPF-86 for Seabrook Station, Unit No. 1 was issued on September 10, 2012 (ML12178A537). This amendment provides permanent application of steam generatortube alternaterepaircriteria,H*.

2. Seabrook Station will to address the potenitial for- cracking of the primary to secondary) pr-essufe boundar-y due to PWSCC of tuabe to tubesheet welds using one of the following to Because alternate repaircriteria has been permanently applied, no plant-specificAMP or alternativeaging management method is required.

1) Perform a one time inspection of a r-epresentative sample of tube to tuibesheet welds in all steam generators to dctcrniine if PWSCC cr-acking is present and, if cr-acking i identified, resolve the condit ion through engineering evalua*,on justiling continued oper-ation or- repair- the oJndit ion, as appropr-iate, and establish an onong moitoring program to per-form r-outine tu be to tubesheet weld inspections for- the rminglife oe the steamn generators, or

2) Perfo*fm an analiytil evaluation showing that the structural interitvy of the steam

/

generator tube to tubeshect inter-face is adequately maintaining the pressur-e bouindary in-the presence of tube to tubesheet weld cracking, and or- by redefining the pr.essur.e boundary ..in hieh the tube to tubesheet weld is no longer included and, therefore, is not requir-ed for reactor coolant pressure bouindary fuinction. The redefinition of the r-eactor coolant pr-essure boundary must be approved by the NRC as paft of a license amendment

.equest.

Based on the above discussion, the following changes have been made to the Seabrook License Renewal Application.

1. In Section A.2.1.10, on page A-10, the second paragraph is revised as follows:

United States Nuclear Regulatory Commission Page 5 of 6 SBK-L-12258 / Enclosure 1 Seabrook Station will-adrfess has addressed the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds. "uing-e, -f*t-he following two option.s: Amendment 131 to the Facility OperatingLicense No. NPF-86for Seabrook Station, Unit No. 1 was issued on September 10, 2012 (ML12178A537). This amendment provides permanent application of steam generator tube alternate repair criteria,Ht.

1) Pefeorm a one time inspeetion of a representative sample of tube to tubesheet welds in all steam gencrators to determine if PWSCC cracling is present and, if cakingi identified, resolve thce ondition theugh enginecringevaluatlin justifying continued epereation or- repair- thc condition, te, and establish an ongoing m itoing proegram to per-form rout~ine tube to tubesheet weld inspections for- the r-emaining life ot the stcam gcnerators, or

2) Perform an analylical evaluation showing that the stmuetural int~egh-ty of thc steam generator tube to tubesheet interface is adequately the presence of tube to tubeshect weld cracking, or-maintaining thethe by redefining pressure boundarn*

pr-essurfe J boundar-y in which the tube to tubesheet weld is no longer- included and, therefore, is not required for reactor coolant pr-essurce boundary frntioein. The redefinition of thc r-ector coolat pr-essurc bouindary muitst bc approeved by the NRC as pai of a lices aedment r-equcst Option 1 or- 2 will bc completed at least 24 months prior-to entering the period of extended

2. In Section A.3, License Renewal Commitment #54 has been revised to designate this action as complete (see Enclosure- 3).

3. In Section B.2. 1.1O, on page B-64, the enhancement previously added under SBK-L- 12123, (Reference 3) has been deleted as follows:

1. hin Section B.2. 1.10, on page B 614, the following Entha-ncement has been add:

Seabrook Station will address the potential for-craeking of the primary to secondar~y pressure boundary due to PWSCC of tube to tubesheet welds using one of the following two options:

1) Per-form a one timne inspection of a repr-esenttative sample of tube to tubesheet wxed i identified, resolve the conditiont thoeugh engineering evaluation j usti6'ing continued operation or- repair- the, condition, as appropriate, and establish an ongoing monefitoriing program to perform r-outine tube to tuibesheet weld inspections for the remfaininig lif of the steamf generators, or

2) Perform an anialytical evaluiation showing that the structural integrity of the steamn generator tube to tubesheet inter-face is adequately maintaining the pressurfe bouindar-y in the presence of tube to tubesheet weld cracking, or- by redefining the pressur boundary in which the tube to tubesheet weld is no longer included and, ther-efore, tis rector coolantA pressurfe boundar-y must be approved by the NRC as paft of a licene amendment request.

United States Nuclear Regulatory Commission Page 6 of 6 SBK-L-12258 / Enclosure I Option .

Enclosure 2 to SBK-L-12258 Clarification of Response to RAI B.1.4-4 c)

United States Nuclear Regulatory Commission Page 2 of 3 SBK-L-12258 / Enclosure 2 In NextEra Energy Seabrook, LLC letter SBK-L-12183 (Reference 6), NextEra responded to NRC requests for additional information related to operating experience. During a telephone conference with the NRC on December 4, 2012, additional information was requested pertaining to the response to RAI B. 1.4-4 c), as to the responsibilities of the License Renewal Engineer, and sharing site operating experience with the rest of the nuclear industry. The response to RAI B. 1.4-4 c) is modified as follows:

RAI B.1.4-4 c) Request Describe guidelines that specifically address reporting of operating experience on age-related degradation and aging management to the industry. In addition, revise the UFSAR supplement to address reporting of plant-specific operating experience related to aging to the industry.

Additional Response Information:

A recent revision to the NextEra procedure for operating experience program screening and responding to incoming operating experience incorporates the guidance of LR-ISG-2011-05 and includes a License Renewal Engineer in the composition of the site operating experience screening team, further assuring age-related degradation and aging management operating experience will be properly reported to the industry.

NextEra Seabrook utilizes the INPO Consolidated Event System (ICES) for sharing of Operating Experience (OE) and lessons learned internally and with other utilities. Reporting is facilitated by a Site OE Coordinator.Involvement of a subject matter expert, such as the License Renewal Engineer,ensures accurateinformation is entered into the ICES Record and reportedto the industry.

The duties of the License Renewal Engineerinclude the following:

1. Prepares and administers the site's license renewal program and implementation project planfollowing issuance of the Renewed License(s).

2. Monitors the status and implementation of site license renewal commitments, on a continuing basis.

3. Ensures Engineering management is informed of program and/or license renewal commitment compliance issues.

4. Provides 10 CFR 54.37(b) input to the UFSAR update, including the coordination of processing newly identifiedSSCs.

5. Maintains cognizance of related industry working groups such as the NEI License Renewal Implementation Working Group.

6. Evaluates new NRC regulatory guidance that may impact implementation activities, such as regulatoryguides, FrequentlyAsked Questions (FAQs), Interim Staff Guidance (ISG),

etc.

7. Reviews applicable Q List (engineeringdata base) revisions.

8. Assisting in the development of initialand continuingprogramowner trainingmaterials.

9. Reviews ARs (Action--Requests) and operating experience for age-related failures or significant degradation of in-scope SSCs, or failures of aging management programs to prevent failures and degradation and initiates changes to site-specific AMPs as appropriate.

United States Nuclear Regulatory Commission Page 3 of 3 SBK-L-12258 / Enclosure 2

10. Performs aging management reviews/assessments, including extent of condition for age-relatedfailures/degradationof in-scope SSCs or AMPs.

The duties of the Site OE Coordinatorinclude thefollowing:

1. Serves as the plant expert on in-house and industry operating experience programs and serves as the site advocatefor the use of operating experience and associatedtools, such as the 1NPO Web site.

2. Acts as the Site OE Champion to ensure effective screening, assignment, disposition, and processingof incoming OE.

3. PerformsICES (INPO ConsolidatedEvent System) data entry or provides input/updates to the Site ICES Coordinator,as required.

4. Supports ICES Coordinator with presentations or updates to the Station MRC (ManagementReview Committee) regardingICES entries and status, as needed.

5. Shares lessons learned with other utilities and NextEra sites and promotes industry-wide safety and reliability,the OE Coordinator:

A. Monitors the condition reporting system to identify potential events that need to be reported to the industry via INPO.

B. Administers the process for identification of internal OE and distributes to other NextEra sites.

C. Ensures accurate and timely information sharing with the INPO Events Analysis Screener.

D. Monitors the disposition of OE distributedto facility organizationsfor evaluation and action, as appropriate.

6. Ensures that periodic station OE Program effectiveness reviews and self-assessments are performed in accordancewith this procedure.

7. Assists plantpersonnelin performingsearches orpreparingreports using ICES data.

Enclosure 3 to SBK-L-12258 LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Completion Status Changes

United States Nuclear Regulatory Commission Page 2 of 12 SBK-L-12258 / Enclosure 3 A.3 LICENSE RENEWAL COMMITMENT LIST UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Program to be implemented prior to the period of extended operation.

Inspection plan to be submitted to An inspection plan for Reactor Vessel Internals will be NRC not later than 2 years after submitted for NRC review and approval. receipt of the renewed license or not less than 24 months prior to the period of extended operation, whichever comes first.

2. Closed-Cycle Cooling Enhance the program to include visual inspection for cracking, Prior to the period of extended Water loss of material and fouling when the in-scope systems are A.2.1.12 operation opened for maintenance.

Inspection of Overhead Enhance the program to monitor general corrosion on the Prior to the period of extended crane and trolley structural components and the effects of wear A.2.1.13 operaton

3. Heavy Load and Light Load (Related to Refseling) on the rails in the rail system. operation Handling Systems Inspection of Overhead Heavy Load and Light Load Prior to the period of extended (Related to Refueling) Enhance the program to list additional cranes for monitoring. A.2.1.13 operation Handling Systems Enhance the program to include an annual air quality test Prior to the period of extended

5. Compressed Air Monitoring requirement for the Diesel Generator compressed air sub A.2.1.14 operation system.

6. Fire Protection Enhance the program to perform visual inspection of A.2.1.15 Prior to the period of extended penetration seals by a fire protection qualified inspector. operation.

Enhance the program to add inspection requirements such as Prior to the period of extended

7. Fire Protection spalling, and loss of material caused by freeze-thaw, chemical A.2.1.15 attack, and reaction with aggregates by qualified inspector. operation.

United States Nuclear Regulatory Commission Page 3 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to include the performance of visual Prior to the period of extended 8, Fire Protection inspection of fire-rated doors by a fire protection qualified A.2.1.15 Priratote inspector. operation.

Enhance the program to include NFPA 25 guidance for "where sprinklers have been in place for 50 years, they shall Prior to the period of extended 9, Fire Water System be replaced or representative samples from one or more A.2.1.16 operation.

sample areas shall be submitted to a recognized testing laboratory for field service testing".

Enhance the program to include the performance of periodic Prior to the period of extended

10. Fire Water System flow testing of the fire water system in accordance with the A.2.1.16 operation.

guidance of NFPA 25.

Enhance the program to include the performance of periodic visual or volumetric inspection of the internal surface of the fire protection system upon each entry to the system for routine or corrective maintenance. These inspections will be documented and trended to determine if a representative Within ten years prior to the period

11. Fire Water System number of inspections have been performed prior to the period A.2.1.16 of extended operation.

of extended operation. If a representative number of inspections have not been performed prior to the period of extended operation, focused inspections will be conducted.

These inspections will be performed within ten years prior to the period of extended operation.

12. Aboveground Steel Tanks Enhance the program to include components and aging effects A.2.1.17 Prior to the period of extended required by the Aboveground Steel Tanks. operation.

Enhance the program to include an ultrasonic inspection and Within ten years prior to the period

13. Aboveground Steel Tanks evaluation of the internal bottom surface of the two Fire A.2.1.17 of extended operation.

Protection Water Storage Tanks.

Enhance program to add requirements to 1) sample and

14. Fuel Oil Chemistry analyze new fuel deliveries for biodiesel prior to offloading to A.2.1.18 Prior to the period of extended the Auxiliary Boiler fuel oil storage tank and 2) periodically operation.

sample stored fuel in the Auxiliary Boiler fuel oil storage tank.

United States Nuclear Regulatory Commission Page 4 of 12 SBK-L-12258 / Enclosure 3 UF'SAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to add requirements to check for the Prior to the period of extended

15. Fuel Oil Chemistry presence of water in the Auxiliary Boiler fuel oil storage tank A.2.1.18 operation.

at least once per quarter and to remove water as necessary.

Enhance the program to require draining, cleaning and Prior to the period of extended

16. Fuel Oil Chemistry inspection of the diesel fire pump fuel oil day tanks on a A.2.1.18 operation.

frequency of at least once every ten years.

Enhance the program to require ultrasonic thickness measurement of the tank bottom during the 10-year draining,

17. Fuel Oil Chemistry cleaning and inspection of the Diesel Generator fuel oil A.2.1.18 Prior to the period of extended storage tanks, Diesel Generator fuel oil day tanks, diesel fire operation.

pump fuel oil day tanks and auxiliary boiler fuel oil storage tank.

Enhance the program to specify that all pulled and tested Prior to the period of extended

18. Reactor Vessel Surveillance capsules, unless discarded before August 31, 2000, are placed A.2.1.19 operation.

in storage.

Enhance the program to specify that if plant operations exceed the limitations or bounds defined by the Reactor Vessel

19. Reactor Vessel Surveillance Surveillance Program, such as operating at a lower cold leg Prior to the period of extended temperature or higher fluence, the impact of plant operation operation.

changes on the extent of Reactor Vessel embrittlement will be evaluated and the NRC will be notified.

Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be withdrawn at an outage in which the

20. Reactor Vessel Surveillance capsule receives a neutron fluence that meets the schedule A.2.1.19 Prior to the period of extended requirements of 10 CFR 50 Appendix H and ASTM E 185-82 operation.

and that bounds the 60-year fluence, and the remaining capsule(s) will be removed from the vessel unless determined to provide meaningful metallurgical data.

Enhance the program to ensure that any capsule removed, 21 Reactor Vessel Surveillance without the intent to test it, is stored in a manner which 1.19 Prior to the period of extended maintains it in a condition which would permit its future use, operation.

including during the period of extended operation.

United States Nuclear Regulatory Commission Page 5 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Within ten years prior to the period

22. One-Time Inspection Implement the One Time Inspection Program. A.2.1.20 of tended prion.

of extended operation.

Implement the Selective Leaching of Materials Program. The Selective Leaching of program will include a one-time inspection of selected Within five years prior to the period

23. Materials components where selective leaching has not been identified A.2.1.21 of extended operation.

and periodic inspections of selected components where selective leaching has been identified.

24. Buried Piping And Tanks Implement the Buried Piping And Tanks Inspection Program. A.2.1.22 Within ten years prior to entering Inspection the period of extended operation One-Time Inspection of Implement the One-Time Inspection of ASME Code Class 1 Within ten years prior to the period

25. ASME Code Class 1 Small Small Bore-Piping Program. A.2.1.23 of extended operation.

Bore-Piping Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects of

26. External Surfaces interest, the refueling outage inspection frequency, the A.2.1.24 Prior to the period of extended Monitoring inspections of opportunity for possible corrosion under operation.

insulation, the training requirements for inspectors and the required periodic reviews to determine program effectiveness.

Inspection of Internal

27. Surfaces in Miscellaneous Implement the Inspection of Internal Surfaces in A.2.1.25 Prior to the period of extended Piping and Ducting Miscellaneous Piping and Ducting Components Program. operation.

Components Enhance the program to add required equipment, lube oil

28. Lubricating Oil Analysis analysis required, sampling frequency, and periodic oil A.2.1.26 operation.

_changes.

Enhance the program to sample the oil for the Reactor Coolant A.2.1.26 Prior to the period of extended

29. Lubricating Oil Analysis pump oil collection tanks. operation.

Enhance the program to require the performance of a one-time

30. Lubricating Oil Analysis ultrasonic thickness measurement of the lower portion of the A.2.1.26 Prior to the period of extended Reactor Coolant pump oil collection tanks prior to the period operation.

of extended operation.

United States Nuclear Regulatory Commission Page 6 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION

31. ASME Section XI, Enhance procedure to include the definition of "Responsible A.2.1.28 Prior to the period of extended Subsection IWL Engineer". operation.

Structures Monitoring Enhance procedure to add the aging effects, additional Prior to the period of extended

32. Program locations, inspection frequency and ultrasonic test A.2.1.31 operation.

requirements.

Structures Monitoring Enhance procedure to include inspection of opportunity when Prior to the period of extended

33. Program planning excavation work that would expose inaccessible A.2.1.31 operation.

concrete.

Electrical Cables and Connections Not Subject to Implement the Electrical Cables and Connections Not Subject Prior to the period of extended

34. 10 CFR 50.49 to 10 CFR 50.49 Environmental Qualification Requirements A.2.1.32 operation.

Environmental Qualification program.

Requirements Electrical Cables and Connections Not Subject to Implement the Electrical Cables and Connections Not Subject

35. En10CFR o 50.49 to 10 CFR 50.49 Environmental Qualification Requirements A.2.1.33 operation.

Environmental Qualification Used in Instrumentation Circuits program.

Requirements Used in Instrumentation Circuits Inaccessible Power Cables Not Subject to 10 CFR Implement the Inaccessible Power Cables Not Subject to 10 Prior to the period of extended

36. 50.49 Environmental CFR 50.49 Environmental Qualification Requirements A.2.1.34 operation.

Qualification Requirements program.

Prior to the period of extended Implement the Metal Enclosed Bus program. A.2.1.35 operaton.

37. Metal Enclosed Bus operation.

Prior to the period of extended

38. Fuse Holders Implement the Fuse Holders program. A.2.1.36 operaton.

operation.

United States Nuclear Regulatory Commission Page 7 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Electrical Cable Connections Not Subject to Implement the Electrical Cable Connections Not Subject to 10 Prior to the period of extended

39. 10 CFR 50.49 CFR 50.49 Environmental Qualification Requirements A.2.1.37 operation.

Environmental Qualification program.

Requirements Prior to the period of extended Implement the 345 KV SF 6 Bus program. A.2.2.1 operaton.

40. 345 KV SF6 Bus operation.

41. Metal Fatigue of Reactor Enhance the program to include additional transients beyond A.2.3.1 Prior to the period of extended Coolant Pressure Boundary those defined in the Technical Specifications and UFSAR. operation.

42. Metal Fatigue of Reactor Enhance the program to implement a software program, to Coolant Pressure Boundary count transients to monitor cumulative usage on selected A.2.3.1 Prior to the period of extended operation.

components.

Pressure -Temperature The updated analyses will be Limits, including Low Seabrook Station will submit updates to the P-T curves and submitted at the appropriate time to

43. Temperature Overpressure LTOP limits to the NRC at the appropriate time to comply A.2.4.1.4 comply with 10 CFR 50 Appendix Temperotion Lim ess with 10 CFR 50 Appendix G. G, Fracture Toughness Protection Limits I I I Requirements.

United States Nuclear Regulatory Commission Page 8 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION NextEra Seabrook will perform a review of design basis ASME Class 1 component fatigue evaluations to determine whether the NUREG/CR-6260-based components that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for the Seabrook plant configuration. If more limiting components are identified, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage. If the limiting location identified consists of nickel alloy, the environmentally-assisted fatigue calculation for nickel alloy will be performed using the rules of NUREG/CR-6909.

(1) Consistent with the Metal Fatigue of Reactor Coolant Pressure Boundary Program Seabrook Station will update the fatigue usage calculations using refined fatigue analyses, if necessary, to determine acceptable CUFs (i.e., less than 1.0) when accounting for the effects of the reactor water Environmentally-Assisted environment. This includes applying the appropriate Fen At least two years prior to entering

44. A.2.4.2.3 Fatigue Analyses (TLAA) factors to valid CUFs determined from an existing fatigue the period of extended operation.

analysis valid for the period of extended operation or from an analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case).

(2) If acceptable CUFs cannot be demonstrated for all the selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, if CUF, including environmental effects is greater than 1.0, then Corrective Actions will be initiated, in accordance with the Metal Fatigue of Reactor Coolant Pressure Boundary Program, B.2.3.1. Corrective Actions will include inspection, repair; or replacement of the affected locations before exceeding a CUF of 1.0 or the effects of fatigue will be managed by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC).

United States Nuclear Regulatory Commission Page 9 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION

45. Number Not Used Protective Coating Enhance the program by designating and qualif~ing an Prior to the period of extended

46. Monitoring and Inspector Coordinator and an Inspection Results Evaluator. A.2.1.38 operation Maintenance Enhance the program by including, "Instruments and Protective Coating Equipment needed for inspection may include, but not be

47. Monitoring and limited to, flashlight, spotlights, marker pen, mirror, A.2.1.38 operation Maintenance measuring tape, magnifier, binoculars, camera with or without wide angle lens, and self sealing polyethylene sample bags."

Protective Coating Enhance the program to include a review of the previous two Prior to the period of extended

48. Monitoring and monitoring reports.A.2.1.38 Maintenance Protective Coating Enhance the program to require that the inspection report is to

49. Monitoring and be evaluated by the responsible evaluation personnel, who is A.2.1.38 Prior to the period of extended Maintenance to prepare a summary of findings and recommendations for operation future surveillance or repair.

Within the next two refueling outages, OR15 or OR16, and

50. ASME Section XI, Perform UT testing of the containment liner plate in the A.2.1.27 repeated at intervals of no more Subsection IWE vicinity of the moisture barrier for loss of material. than five refueling outages

51. Number Not Used ASME Section XI, Implement measures to maintain the exterior surface of the

52. Subsection W Containment Structure, from elevation -30 feet to +20 feet, in A.2.1.28 Ongoing By Deccmbcr 31, 2012 a dewatered state.

Replace the spare reactor head closure stud(s) manufactured Prior to the period of extended

53. Reactor Head Closure Studs from the bar that has a yield strength > 150 ksi with ones that A.2.1.3 operation.

do not exceed 150 ksi.

United States Nuclear Regulatory Commission Page 10 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION NextEra will address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options:

1) Perform a one-time inspection of a representative sample of tube-to-tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine At least 24 monthsprior-t entering Steam Generator Tube tube-to-tubesheet weld inspections for the remaining life of the A.2.1.10 th".. riod cf extended operatien.

54. Integrity steam generators, or A... 1 .....

2) Perform an analytical evaluation showing that the structural Complete integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence of tube-to-tubesheet weld cracking, or redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function. The redefinition of the rector coolant pressure boundary must be approved by the NRC as part of a license amendment request.

55. Steam Generator Tube Seabrook will perform an inspection of each steam generator A.2.1.10 Within five years prior to entering Integrity to assess the condition of the divider plate assembly. the period of extended operation.

Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the period of

56. Water System Guideline operating ranges and Action Level values for A.2.1.12 extended operation.

hydrazine and sulfates.

57 Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the period of Water System

57. Guideline operating ranges and Action Level values for Diesel A.2.1.12 extended operation.

Generator Cooling Water Jacket pH.

Update Technical Requirement Program 5.1, (Diesel Fuel Oil Prior to the period of extended

58. Fuel Oil Chemistry Testing Program) ASTM standards to ASTM D2709-96 and A.2.1.18 operation.

ASTM D4057-95 required by the GALL XI.M30 Rev I Nickel Alloy Nozzles and The Nickel Alloy Aging Nozzles and Penetrations program Prior to the period of extended

59. ni oyn will implement applicable Bulletins, Generic Letters, and staff A.2.2.3 operation.

accepted industry guidelines.

ffi . . 16 United States Nuclear Regulatory Commission Page 11 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Implement the design change replacing the buried Auxiliary A.2.r1.22 Boiler supply piping with a pipe-within-pipe configuration A.2.1.22 extend erin.

60. Buried Piping and Tanks Inspection with leak detection capability. extended operation.

61. Compressed Air Monitoring Replace the flexible hoses associated with the Diesel A.2.1.14 Within ten years prior to entering Program Generator air compressors on a frequency of every 10 years. the period of extended operation.

Enhance the program to include a statement that sampling Prior to the period of extended

62. Water Chemistry frequencies are increased when chemistry action levels are A.2.1.2 operation.

exceeded.

Ensure that the quarterly CVCS Charging Pump testing is continued during the PEO. Additionally, add a precaution to

63. Flow Induced Erosion the test procedure to state that an increase in the CVCS N/A Prior to the period of extended Charging Pump mini flow above the acceptance criteria may operation be indicative of erosion of the mini flow orifice as described in LER 50-275/94-023.

Soil analysis shall be performed prior to entering the period of extended operation to determine the corrosivity of the soil in

64. Buried Piping and Tanks the vicinity of non-cathodically protected steel pipe within the A.2.1.22 Prior to entering the period of scope of this program. If the initial analysis shows the soil to extended operation.

Inspection be non-corrosive, this analysis will be re-performed every ten years thereafter.

Implement measures to ensure that the movable incore Prior to entering the period of

65. Flux Thimble Tube detectors are not returned to service during the period of N/A extended operation extended operation.

66. Number Not Used Perform one shallow core bore in an area that was

67. Structures Monitoring continuously wetted from borated water to be examined for A.2.1.31 No later than December 31, 2015 Program concrete degradation and also expose rebar to detect any degradation such as loss of material.

Structures Monitoring Perform sampling at the leakoff collection points for

68. Program chlorides, sulfates, pH and iron once every three months. A.2.1.31 Starting January 2014

United States Nuclear Regulatory Commission Page 12 of 12 SBK-L-12258 / Enclosure 3 UFSAR PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Replace the Diesel Generator Heat Exchanger Plastisol PVC Prior to the period of extended lined Service Water piping with piping fabricated from A.2.1.11 operaton.

69, Open-Cycle Cooling Water System AL6XN material. operation.

Inspect the piping downstream of CC-V-444 and CC-V-446 70, Closed-Cycle Cooling Water to determine whether the loss of material due to cavitation A.2.1.12 Within ten years prior to the period System induced erosion has been eliminated or whether this remains of extended operation.

an issue in the primary component cooling water system.

7 Alkali-Silica Reaction Implement the Alkali-Silica Reaction (ASR) Monitoring Prior to entering the period of (ASR) Monitoring Program Program extended operation.